IAEA-IWG-NPPCI-95/12 . LIMITED DISTRIBUTION WORKING MATERIAL INSTRUMENTATION AND EQUIPMENT FOR MONITORING AND CONTROLLING NPP POST-ACCIDENT SITUATIONS Proceedings of a Specialists9 Meeting Organized by the International Atomic Energy Agency in co-operation with Research Institute of Atomic Reactors (RIAR) and held in Dimitrovgrad, Russian Federation, 12-15 September 1995 Reproduced by the IAEA Vienna, Austria, 1995 NOTE The material in this document has been supplied by the authors and has not been edited by the IAEA. The views expressed remain the responsibility of the named authors and do not necessarily reflect those of the government(s) of the designating Member State(s). In particular, neither the IAEA nor any other organization or body sponsoring this meeting can be held responsible for any material reproduced in this document. We regret that some of the pages in the microfiche copy of this report may not be up to the proper legibility standards, even though the best possible copy was used for preparing the master fiche Poor print: Sc J a. j FOREWORD Accidents in nuclear power plants are postulated or un-postulated events that must not lead to a significant release of radioactive material to the environment. Accident management involves the actions taken by the nuclear power plant staff during the course of an accident to: prevent core damage, terminate progress of core damage, maintain containment integrity and minimize off-site releases. Traditionally, nuclear utilities have relied on event-oriented emergency procedures, that is, planned, manually-controlled actions following reactor trip and other safety actions that are automatically performed. These procedures are based on design basis accident scenarios that are evaluated in the plant safety analysis. However, accident evolution may follow a route other than that predicted. Furthermore, combination of events and faults may be such that the accident cannot be clearly identified. To confront these problems, safety-function oriented emergency procedures have been proposed. In this approach there is less emphasis on accident characterization and more on restoring the plant to safe conditions based on the values of representative plant variables. Due to the difficulties in determining cause and type of accident, fully automated safety systems are designed to mitigate the effects of postulated accidents. Operators need indications in the control room to verify that these safety systems succeeded in performing their intended safety functions. For example, measurements of neutron flux, control (or shutoff) rod position will reveal whether the shutdown system actuated successfully. Ideally, accident and post-accident management should be conducted in the main control room by a special team composed operators and safety engineers. The team's ability to mitigate the consequences of the accident will strongly depend on the adequacy and reliability of the monitoring system. Accident and post-accident monitoring requires special instrumentation that is independent of normal power plant instrumentation but qualified to survive the severest conditions associated with design-basis accidents. Normal power instrumentation may perform satisfactorily in the early phases of the accident but will gradually become unreliable and off-scale as the accident progresses and the measured quantities deviate significantly from the normal operating conditions. Instrumentation is also provided to assist operator actions involving safety-related systems that need to be manually started. Accident-monitoring instrumentation, and the variables being monitored, should be similar to those used in normal power operation. This will assure that the operator is familiar with the characteristics and behavior of the instruments. ReUabiHty of the sensor signals is another serious concern because computer processing of the input data may sometimes mask sensor failures. Therefore some method of on-line signal validation is needed. Possible techniques include: consistency checks between redundant channels, noise detection for signal changes and process empirical modelling, i.e., the use of simple mathematical models that correlate the variables of the physical process. 1 Following the Three Mile Island accident several studies recommended the use of computerized aids to assist operators in monitoring plant safety-related information, correcting abnormal conditions and providing feedback to corrective actions. The idea has been embraced by nuclear utilities and computerized safety panels were installed in the main control rooms of many nuclear power plants. The safety panel may perform numerous functions including: monitoring the main safety variables, displaying the chronology of faults and the status of the safety actions and assisting the operator in identifying the procedure to be followed. The use of digital technology in the safety panel has considerable advantages over conventional analog displays, among which are: lower costs, reliability, the ability to display information in different ways, use of mathematical models to interpret the ongoing phenomena, flexibility to modify the software and facility to interface with other digital devices. It is essential the operators participate in the development of these monitormg/diagnostics systems from the start. For example, operators can conduct preliminary tests on full scope simulators and provide the system designer with important feedback on system weaknesses and additional operational needs. The Specialists' Meeting on "Instrumentation and Equipment for Monitoring and Controlling NPP Post-Accident Situations" was organized by the IAEA in co-operation with the Research Institute of Atomic Reactors (RIAR), Dimitrovgrad, Russian Federation. The meeting brought together experts on power plant operation with experts on application of today's instrumentation and control technology. In this way, a match can be made between those knowing the industry needs and requirements and those knowing the potentials of the technology. The objectives of the Specialist's Meeting were: To provide an international forum for presentation and discussion on experience with mstrumentation and equipment for monitoring and controlling NPP post-accident situations. To share experience on the design and improvements in the subject area.. To identify and describe advanced features for safety and post-accident management improvements. In order to facilitate a structured discussion and not to omit important problem areas, papers on the following subjects were considered to be within the scope of the Specialist's Meeting: Analysis of existing post-accident instrumentation (in terms of range, qualification and redundancy) to support post-accident management. 2 3 Requirements for post accident mstromentation and equipment - operating experience, regulations, standards. Experience in upgrading of the existing instrumentation. Instrumentation and equipment for emergency management systems (emergency rooms and national centers). Post-accident computerized aids to assist operators. Post-accident operating procedures. R&D in post-accident instrumentation, future trends. The present volume contains: (1) the papers presented by national delegates, (2) programme of the meeting, (3) description of the RIAR research facilities, and (4) list of participants. s CONTENTS PAPER PRESENTED AT THE MEETING LI Improvement in Post-Accident Instrumentation for Spanish Nuclear Power Plants Rafael Cid, Spain 1.2 Safety Parameter Display System for Kalinin NPP V.l. Andreev,KN. Videneev, NN. Davidenko, G.I. S haft an, V.G. Dounaev and V.T. Neboyan, Russian Fed., J-C. Tissot, D. Joonekindt, France 1.3 Development of Transmitter with Hybrid-IC for Post-Accident Monitoring Instrumentation T. Ikeuchi and S. Watanabe, Japan 1.4 Relative Different Energy Neutron Radiometry in Reactors for Preventing of Accidents Caused by Uncontrolled Reactivity Variations S. V. Volkov, Russian Federation 1.5 A Stochastic Approach to Accident Identification in Nuclear Power Plants Kee-Choon Kwon, Soon-Ja Song, Won-Man Park, Joe-Chang Park and Chang- Shik Ham, Rep. of Korea 1.6 Accident Monitoring in Ventilation Stack V.Kapisovsky, V.Zbiejczukovâ, F.Gâbris, G.Belan, JZentan, J.Bukovjan, I.Rehak, Slovak Rep. 1.7 Experience with Neutron Flux Monitoring Systems Qualified for Post-Accident Monitoring KG. S Imgars and IF. Miller, USA. 1.8 Basic data of Emergency Response Centre O. Jenicek, Czech Republic 1.9 Applied Software of the Emergency Recording System for Reactor Facility Parameters under the Minor Statistics Conditions V.B. Ivanov, A.F. Grachev, O.M. Kinsky, RS.Makin, A.I. Ochrimenko, L.I. Demidov, V.l. Karpjuk, V.K. Afonin, RG. Iakanderov, Russian Fed. 1.10 Design Implementation of the Post-Accident Monitoring (РАМ) System for Wolsong NPP Units 2,3 & 4 in Korea San-Joon Han, Rep. of Korea 1.11 Electrical Characteristics of Nuclear Power Plant Cables at Accident Simulation V.l. Rogov, V.N. Ulimov, N.I. Filatov, V.S. Shestacov, Russian Fed. 1 6 1.12 Research Istitute of Atomic Reactors: General Presentation V.B. Ivanov, Russian Fed. 1.13 Creating Requirements for Control of Post-Accident Situations at NPP V.S. Dickarev, V.S. Ionov, Russian Fed. 1.14 Principles of Elaboration and Creation of Information-Analytical System "RI Operation Safety with SSC RIAR Research Reactors" V.B. Ivanov, A.F. Grachev, O.M. Kinsky, RS. Makin, A.I. Ochrimenko, L.I Demidov, V.l. Karpjuk, V.K. Afonin, RG. Iskanderov,Russian Fed. 1.15 Radiological Consequences